The integral effects facilities are scaled down and contain
atypicalities; therefore plant measured data are very important for code
assessment. The plant measured data come mostly from startup tests and
operational events. Several such tests and events have been analyzed in
the past [1-7]. Nevertheless, when an operational event occurred, not
all plant data may be measured, which are important for interpretation
of the events and physical understanding. Also, for some measured
parameters part of history may be missing. In such a case well-simulated
transient progression can help in reconstructing the plant response. The
purpose of this study was twofold: on one hand to assess the latest
version of RELAP5/MOD3.3 Patch 05 and on the other hand to better
understand the plant response to the event with the reactor trip and
safety injection signal actuation. The International Atomic Energy
Agency (IAEA) specific safety guide number SSG-2 [8] (now in revision)
states that "accident analyses may be used as a tool for obtaining
a full understanding of events that occur during the operation of
nuclear power plants and should form an integral part of the feedback
from operating experience." According to IAEA SSG-2 [8] the
objective to analyze the operational events may be also to provide
additional information on the time dependence of the values of
parameters that are not directly available using the plant
instrumentation, to check whether the plant operators and plant systems
performed as intended and to validate and adjust the models in the
computer codes that are used for analyses. In the present work, the
RELAP5/MOD3.3 analysis of the abnormal event, which occurred on March
23, 2011, in Krsko nuclear power plant (NPP), is presented. This event
has been initially analyzed by earlier Patch 04 version of RELAP5/MOD3.3
computer code [9]. First the methods are described including abnormal
event, standard RELAP5 input model of Krsko NPP, and scenarios
simulated. The RELAP5/MOD3.3 simulated results are then presented and
compared to measured data that were available, followed by the
discussion. At the end, the conclusions are given.

2. Methodology Description

First abnormal event is described for the selected plant. Then, the
RELAP5/MOD3.3 Patch 05 [10] used for calculations and the input model
for selected pressurized water reactor (PWR) are described. Finally, the
simulated scenarios are described.

2.1. Abnormal Event Description. Krsko NPP is a two-loop PWR,
Westinghouse type. Table 1 shows the main sequence of events. An
abnormal event occurred on March 23, 2011, at 10:27. The initiating
event was spurious activation of 400 kV bus differential protection in
the NPP's 400 kV switchyard. Therefore the circuit breakers in the
bays were disconnected, which caused loss of offsite load [11]. The
turbine runback started to reduce the power. The turbine control system
and steam dump system were not successful to stabilize the power and
main steam line pressure. Approximately two minutes after the initiating
event the low steam line pressure actuated the safety injection, which
caused the reactor and turbine trip. The diesel generators automatically
started on SI signal as designed (up to 10 s) and safety systems started
according to SI sequence (high pressure and low pressure safety
injection pump with 5 s delay on SI signal and auxiliary feedwater pump
with 20 s delay on SI signal). No measured start times were available,
but it was reported that diesel generators and SI sequence performed as
designed. Also, a loss of power supply occurred to safety buses and
nonsafety buses because the generator load switch has been automatically
disconnected. This resulted in the reactor coolant pumps stop. At 10:55,
an "unusual event" was declared according to the emergency
response procedure for loss of power to 6.3 kV bus, which occurred at
event occurrence. At the time of event declaration, the power has been
supplied to safety buses through the auxiliary transformer. Later both
diesel generators were manually shut down and placed in standby mode.
When plant was safely shut down (bringing down to "hot
standby" mode), the "unusual event" was ended.

The measured data received were recorded in the first four minutes
of event. The measured parameters were available for primary and
secondary side. The primary side parameters were nuclear power, rod
position, pressurizer pressure, hot leg temperatures, cold leg
temperatures, average reactor coolant system (RCS) temperatures,
reference temperature, delta temperatures, overpower, and
overtemperature delta temperature. The secondary side parameters were
turbine first-stage impulse pressure, steam generator pressures and
water levels, steam and feedwater flows, turbine-generator rotor speed,
and turbine governor valves position. The reported operator action was
switching the feedwater flow control from automatic to manual mode with
no further details. No information on steam demand, steam dump open
logic signals, and steam dump flow was given.

2.2. RELAP5 Computer Code Description. The RELAP5/ MOD3.3 is
best-estimate thermal hydraulic computer code delivered through Code
Applications and Maintenance Program (CAMP) of U.S. Nuclear Regulatory
Commission (NRC). It has been developed for best-estimate transient
simulation of light water reactor coolant systems during postulated
accidents such as loss of coolant, anticipated transients without scram
(ATWS), and operational transients such as loss of feedwater, loss of
offsite power, station blackout, and turbine trip. It can be used for
simulating phenomena like core uncovery and core heatup up to the
temperature of significant oxidation of the cladding and hydrogen
production at around 1200[degrees]C to 1500[degrees]C. For more
information on the RELAP5/MOD3.3 Patch 05 code, the reader can refer to
[10].

2.3. Krsko Input Model for RELAP5. Krsko NPP base RELAP5 input
model has been used for simulations. This input model has been used
first for reference calculations for Krsko full scope simulator
verification [12, 13]. The analyses were performed for uprated
conditions (2000 MWt) with the new steam generators (SGs). The power
uprate and steam generators replacement has been done in 2000. Later the
RELAP5 input model has been validated by plant transients, for example,
[2,3]. The input model represented by SNAP (see Figure 1) consists of
304 hydraulic components and 108 heat structures. For more details, the
reader can refer to [9].

The primary side is modelled by the reactor pressure vessel (RPV)
and two primary loops (loops 1 and 2) including reactor coolant pump
(RCP) and U-tubes of steam generator. To the primary loop number 1, the
pressurizer (PRZ) vessel is connected through surge line (SL). At the
top of pressurizer, the pressurizer spray lines, two pressurizer power
operated relief valves (PORVs), and two pressurizer safety valves are
connected. Active high pressure safety injection (HPSI) and low pressure
safety injection (LPSI) pumps, and accumulators (ACCs) represent the
emergency core cooling system.

The secondary side consists of the steam generators with main
feedwater (MFW) and auxiliary feedwater (AFW) system. The secondary side
is modelled (using the logic) up to turbine valve. The MFW and AFW pumps
are modelled as time dependent junctions and turbine as time dependent
volume.

Control variables and general tables are mainly used to model the
control systems. Rod control system has been modelled for point
kinetics. For maintaining the pressure and level in the primary system,
pressurizer pressure and level control are modelled. Another control
system modelled is steam dump system (see Figure 2), which is important
for transient originated on the secondary side. The steam dump control
system consists of turbine trip, loss of load, and steam pressure
controllers. There are four groups of valves, labelled "a,"
"b," "c," and "d." In case of fast opening
the bistables open groups "a," "b," "c,"
and "d" on HI-1 (high 1) demand (20% capacity), on HI-2 demand
(30% capacity), on HI-3 demand (20% capacity), and on HI-4 demand (30%
capacity), respectively.

2.4. Initial and Boundary Conditions Used in RELAP5 Input Model.
The initial conditions used in the standard RELAP5 input model were
compared to plant measured data (see Table 2). Because of good agreement
between the plant measured data and RELAP5 values of initial conditions,
there was no need to perform additional steady-state calculation.

The boundary conditions used for simulation are shown in Figure 3.
For modelling the loss of external load, the artificial turbine control
was used and the turbine power was given as boundary condition. Namely,
the turbine power is a function of measured turbine first-stage pressure
(i.e., impulse pressure), which is shown in Figure 3(a). The turbine
external power represents the input into rod control and steam dump
control system.

As there was no information on manual operation of main feedwater
system, the measured feedwater flow shown in Figure 3(b) was used as
boundary condition.

Because the secondary side is modelled up to the turbine and no
turbine overspeed controller is modelled, the steam flow was input as
boundary condition. The total steam flow was obtained by summing the
measured steam flows in each steam line (see Figure 3(c)).

Finally, turbine-generator frequency in the plant influences the
alternate current (AC) frequency. Again, due to missing turbine model,
the influence of AC frequency (see Figure 3(d)) was inputted by reactor
coolant pump rotor speed as boundary condition. Namely, reactor coolant
pump flow depends on the rotor speed. It should be also noted that the
main target of the automatic turbine control was to maintain the
frequency at 50 Hz (speed 1500 rpm).

2.5. Description of Selected Scenarios. The analysis of abnormal
event was performed in two steps. First we focused on the first part of
transient till SI injection signal generation. Namely, only for
scenarios in which SI signal was generated, it was meaningful to
simulate the transient for the duration of measured data, that is, 4
minutes.

Table 3 shows scenarios selected for sensitivity analysis in the
first step. Considered were also scenarios with operator action and
plant responses not specified in the IRS 8300 report [11]. For example,
the emergency boration was not reported in the IRS 8300 report [11]. We
found the information on emergency boration in the internal plant report
dealing with analysis of the event and scenario "w/o boration"
was selected. Regarding the pressurizer PORV, the plant measured data
show that pressurizer rate sensitive PORV did not open. The way to
follow the plant response was therefore to disable pressurizer rate
sensitive PORV opening. A case with not disabling pressurizer rate
sensitive PORV has been also selected to see how plant input model
behaves in such case (scenario "With PRZ PORV"). The turbine
frequency change influences the AC frequency and the AC frequency
influences the reactor coolant pump operation. The sensitivity run not
considering AC frequency influence has been also performed (scenario
"w/o frequency"). Further, in the base case simulation the
boration was considered for the reactor power tuning. A sensitivity run
without considering boration has been performed (scenario "w/o
boration"), because no information on the emergency boration start
and flowrate of boron injection was available. The scenario with main
feedwater system in automatic mode of operation was performed (scenario
"w/o operator FW") to verify the hypothesis that operator
manual main feedwater control did not help in the secondary pressure
stabilization.

In the second step (simulating 4 minutes of transient) the base
case scenario from Table 3 has been selected. Based on the analysis of
measured data, additional scenario with assuming some steam release
after SI signal generation has been performed. Namely, it is not logical
that after steam line isolation on SI signal the measured steam flow at
the exit of steam generators is nonzero (the validity sign for measured
values was true), if there is no steam flow. Therefore, it may be
assumed that the steam was released either through the SG PORVs or steam
dump and scenario simulated was labelled "With SG PORV."

Because the steam dump flow and turbine flow were not measured, in
the simulations the total steam flow was used as boundary condition. The
purpose was to show that if the reactor and turbine power are balanced,
the RELAP5 plant model is good representative of the plant response.
Turbine power is represented by steam flow to the turbine. To get this
flow, a total steam dump flow was first determined from measured highest
RCS temperature (auctioneered) and the reference temperature (Tref)
signals (see Figure 6(a)), which have been sent through steam dump load
rejection controller. Because RELAP5 computer code is limited by the
number of data points for input signals, the TRACE computer code has
been used to model steam dump load rejection controller (see Figure 4).
The output from the steam dump controller was the temperature error of
the lead-lag compensated input signals, which determine the valve
opening (see Figure 2), and the valve opening determines the steam dump
flow. The difference between the total steam flow and total steam dump
flow is the turbine flow, which has been used as boundary condition in
scenario labelled "With TB flow."

The TRACE model for low steam line pressure signal compensation,
shown in Figure 5, has been used to calculate the lead-lag compensated
signal for low steam line pressure signals. Each steam line has three
channels per steam line. To generate the SI signal, 2 out of 3 lead-lag
compensated low steam line pressures are needed. Knowing the values of
compensated steam line pressure signal helps in understanding the
transient.

3. Results

3.1. Information Obtained from Analysis of Plant Measured Data. The
input temperature signals into steam dump controller and their lead-lag
compensated values are shown in Figure 6(a). The lead-lag compensated
temperature error signal, which is used for load rejection input signal,
is shown in Figure 6(b). The information available to operators is the
temperature difference signal between RCS average and reference
temperature. One may see that, after first 40 seconds, the lead-lag
compensated temperature error signal and temperature difference signal
agree well. The information on the turbine governor valve position
indication is shown in Figure 6(c). One may see that after the transient
start all four governor valves were closed to prevent overspeed of
turbine-generator. Once the speed of turbine-generator started to drop
(see Figure 3(c)), the first two governor valves partially reopen, but
after full steam dump demand (see Figure 6(b)) they started to close
again after 17 seconds. When steam dump group "d" valves fully
close around 40 s, the first two governor valves open again. Then the
position of governor valves is rather constant until 90 s. Five seconds
later both the first two governor valves and steam dump group
"c" started to close, and another five seconds later also
steam dump group "b" valves started to close. Due to
overheating, the secondary side pressure increases. Turbine control
reacted by reopening the valves. The steam dump demand reopens group
"b" valves. The secondary pressure dropped and due to the
lead-lag compensated pressure signal the SI signal was actuated as shown
in Figure 6(d). One may see that secondary pressure trend is far from
dropping below low pressure setpoint of 4.928 MPa which would be
indication of steam line break.

Table 4 shows the minimum values of low pressure steam line
pressure compensated signals predicted by TRACE model of steam dump
control system at around 10 s. It maybe seen that already at this time
the plant was close to SI signal actuation (i.e., minimum value below
4.928 MPa) which was prevented due to the closure of all turbine
governor valves. Later at 95 s, as already mentioned, the first two
governor valves started to close again, and soon after 95 s also the
steam dump group "b" valves started to close. This is in
agreement with steam flow shown in Figure 3(c). Closure of the valves
resulted in the secondary pressure increase. Both turbine control and
steam dump system therefore open the valves, but this time the lead-lag
compensated steam line pressure dropped below setpoint and SI signal was
actuated.

The lead constant has significant influence on compensated low
steam line pressure (i.e., rate of pressure increase and drop).
Therefore, the lead constant was parametrically varied and results are
shown in Figure 7. Using lead constant 65 s in steam dump controller
would cause SI signal actuation at around 10 s. On the other hand,
smaller lead constant would prevent SI signal actuation in the first two
minutes.

3.2. Simulated Results till Plant SI Signal Actuation. The results
are shown in Figures 8-10. The scenarios listed in Table 3 are compared.
Base case calculation (labelled "Base") is performed
considering both borations, changing AC frequency and disabling rate
sensitive PRZ PORV number 2. Figure 8(a) shows that when not considering
boration (labelled "w/o boration") the core power
significantly deviates from the measured data. Therefore also all other
important parameters deviate, as shown in Figures 8-10. The pressurizer
pressure is higher (see Figure 8(b)), steam generator pressures are
higher (see Figures 10(a) and 10(b)), and therefore low steam line
pressure signals (see Figures 10(e) and 10(f)) are higher, not resulting
in actuating the SI signal. The AC frequency change influencing the RCS
flow increases the power (see Figure 8(a)) in first few seconds (not in
the case "w/o frequency"). The power increase caused
pressurizer pressure and steam generator pressures increase. According
to the report [11] the PRZ PORV no. 1 and 2 did not open in first few
seconds and this is in agreement with measured pressurizer pressure
signal shown in Figure 8(b). The consideration of rate sensitive PRZ
PORV in the calculation mainly influences the pressurizer pressure in
the initial period of 13, but it is important for deeper understanding
of the transient. The PRZ PORV no. 1 did not open in the simulation,
while PRZ PORV no. 2 opened in simulation (case "With PRZ
PORV") and by this prevented the significant pressure increase in
the initial period (like in "w/o frequency case"). Figure 9
shows the hot leg, cold leg and average temperatures. They are in
agreement with measured data. The exception is "w/o boration"
and partly "w/o operator FW" case. For the latter case this is
expected because of different control than in the case of manual
operator control. Figure 10 shows the secondary side plant parameters.
They are also in good agreement for all cases except "w/o
boration" and partly "w/o operator FW" case. Figures
10(e) and 10(f) show low steam line compensated pressure no. 1 and 2,
respectively, how they dropped below the setpoint for SI signal
actuation.

3.3. Simulated Results till the Measurement End Time. The results
for duration till the measurement end time are shown in Figures 11 and
12. Two new scenarios have been studied. First was a scenario simulating
some steam flow after SI signal actuation (labelled "With SG
PORV"). The SI signal actuates the main steam line isolation, but
measured steam slow at the exit of steam generators is nonzero (see
Figure 3(c)). If main steam line isolation occurred, the steam may be
discharged through the SG PORVs. If not, the steam may be discharged to
steam dump. The influence of steam flow on the steam generator pressures
is shown in Figures 11(c) and 11(d) and on the cold leg temperatures
shown in Figures 12(a) and 12(b). The simulation results using turbine
flow instead of total steam flow as boundary condition and steam dump
model ("With TB flow" case) are quite similar to
"Base" simulation results. This finding suggests that turbine
flow was correctly predicted.

3.4. Discussion of Results. The results suggest that available
plant information on systems performance is crucial for the whole
transient duration simulation. If not available, the calculated plant
response may be different comparing to the measured data. Also it was
demonstrated that some measured data maybe used to reproduce some
missing information. For example, the steam dump demand signal has been
reproduced, which enables calculation of turbine flow. Another example
is calculation of compensated low steam line pressure signals from steam
line pressure signals. Namely, the safety injection signal is actuated
on compensated signal. With such information the plant transient could
be better understood, especially the operation of turbine valve control
and steam dump system. Also, the hypothesis that operator manual main
feedwater control did not help in secondary pressure stabilization has
been confirmed by simulation. Finally, one important finding seems to be
as follows: almost simultaneous request of two independent controllers
(turbine control and steam dump load rejection controller, resp.)
besides fast lead constant for pressure compensation may be the reason
for SI signal actuation due to fast rate of compensated pressure change.
Also, the main target of automatic turbine control was to control the
speed of the turbine-generator and not to stabilize the steam line
pressure.

4. Conclusions

The abnormal event with loss of external load, followed by safety
injection actuation and reactor trip has been simulated with the latest
RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code.
Comparison between calculated data and plant measured data suggests that
RELAP5 code is appropriate for simulating such abnormal events. It was
demonstrated that certain measured data may be used to reproduce some
missing information. The steam dump demand signal has been reproduced
from measured average and reference temperature signals, which enables
calculation of turbine flow. This further explains operation of the
automatic turbine control besides the steam dump system operation. The
hypothesis that operator manual main feed-water control did not help in
secondary pressure stabilization has been confirmed by simulation. It
has been shown how the steam generator pressure change and lead constant
influence the low steam pressure compensated signal designed to actuate
the safety injection signal. It can be concluded that simulation tools
like RELAP5/MOD3.3 can provide valuable insights into the plant
response.

https://doi.org/10.1155/2018/6964946

Conflicts of Interest

The authors declare that there are no conflicts of interest
regarding the publication of this paper.

[11] International Incident Reporting System (IRS), Reactor Trip
and Actuation of Safety Injection System Caused by the Spurious
Activation of Bus Protection in 400 KV Switchyard, International
Incident Reporting System (IRS), 2012.